A first look at LOCAs in the SBWR using RELAP5/MOD3

Conference · Wed Jan 01 00:00:00 EST 1992

OSTI ID:6639079

Ghan, L S; Shaw, R A; Kullberg, C M

The General Electric Company (GE) is designing an advanced light-water reactor, the Simplified Boiling Water Reactor (SBWR), that utilizes passive safety concepts. The SBWR reactor coolant system will operate on natural circulation with decay heat removal and emergency core coolant injection being provided by passive, gravity-driven systems. The Idaho National Engineering Laboratory has developed an input model of the SBWR for the RELAP5/MOD3 thermal-hydraulic safety analysis code. Preliminary calculations have been performed to simulate three loss-of-coolant accidents: (1) a main steam line break, (2) spurious opening of one automatic depressurization valve, and (3) the rupture of the bottom drain line. Results from these three calculations were, in general, intuitively reasonable. The analyses revealed that the input model, which was created with preliminary design data, needs to be updated to reflect the current SBWR design. Nodalization of certain regions will also need to be improved. The results of the main steam line break calculation were compared to a similar TRACG calculation presented in GE's Standard Safety Analysis Report. Comparisons of the preliminary RELAP5/MOD3 results to TRACG results indicated good qualitative agreement.

• Research Organization: EG and G Idaho, Inc., Idaho Falls, ID (United States)
• Sponsoring Organization: USNRC; Nuclear Regulatory Commission, Washington, DC (United States)
• DOE Contract Number: AC07-76ID01570
• OSTI ID: 6639079
• Report Number(s): EGG-M-92511; CONF-9210204-4; ON: DE93005252
• Resource Relation: Conference: Water reactor safety information meeting, Washington, DC (United States), 21-23 Oct 1992
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
BWR TYPE REACTORS
LOSS OF COOLANT
COMPUTER CALCULATIONS
COMPUTERIZED SIMULATION
HEAT TRANSFER
HYDRAULICS
R CODES
REACTOR COOLING SYSTEMS
REACTOR SAFETY
T CODES
ACCIDENTS
COMPUTER CODES
COOLING SYSTEMS
ENERGY TRANSFER
ENRICHED URANIUM REACTORS
FLUID MECHANICS
MECHANICS
POWER REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTORS
SAFETY
SIMULATION
THERMAL REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210100 - Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled